1. FIELD OF THE INVENTION
The present invention relates to a novel zirconium-based alloy and, more particularly, to a zirconium-based alloy which is suitable for use as a material of fuel cladding tubes in a nuclear reactor, having superior corrosion resistance to withstand the use at high degree of burn-up of the fuel in the nuclear reactor. The invention is concerned also with a nuclear fuel rod having a cladding tube made of the zirconium-based alloy, as well as a nuclear fuel assembly having such fuel rods.
2. DESCRIPTION OF THE PRIOR ART
Among various known zircaloys, most commonly used as the material of a nuclear fuel cladding tube are zircaloy-2 (Sn: 1.20-1.70 wt %, Fe: 0.07-0.20 wt %, Cr: 0.05-0.15 wt %, Ni: 0.03-0.08 wt %, O: 900-1500 ppm and the balance substantially Zr, where (Fe+Cr+Ni): 0.16-0.24 wt %), and zircaloy-4 (Sn: 1.20-1.70 wt %, Fe: 0.18-0.24 wt %, Ni: 0.007 wt % or less, O: 900-1500 ppm, and the balance substantially Zr, where (Fe+Cr): 0.28-0.37 wt %).
The history of development of these zircaloys is described in detail in an article in ASTM, STP No. 368 (1963), pages 3-17. This article also introduces various other zircaloys such as zircaloy-1 (Zr-2.5 wt % Sn), zircaloy-3A (Zr-0.25 wt % Sn-0.25wt % Fe), zircaloy-3B (Zr0.5 wt % Sn-0.4 wt % Fe), zircaloy-3C (Zr-0.5 wt % Sn-0.2 wt % Fe-0.2 wt % Ni), and zircaloy-2 (Sn: 1.20-1.70 wt %, Fe: 0.12-0.18 wt %, Cr: 0.05-0.15 wt %, Ni: 0.007 wt % or less).
These zircaloys other than the zircaloy-2 and zircaloy-4 suffer from the following disadvantages.
The zircaloy-1, which does not contain Fe, Cr and Ni, show only a low level of corrosion resistance. The zircaloys-3A-3C are intended for higher producibility through reduction of the Sn content, as well as for higher corrosion resistance through increasing the Fe and Ni contents. These zircaloys-3A-3C, however, show a low level of strength, that is, about 75% of that exhibited by the zircaloy-2. A Ni-free zircaloy-2 show only small corrosion resistance in 510.degree. C. steam, due to elimination of Ni content. The zircaloy-4 is an alloy which is obtained by increasing the Fe content in the Ni-free zircaloy-2. This alloy, however, has to have a large Fe content due to the elimination of Ni content, with the result that the neutron absorption cross section is increased undesirably.
According to the article mentioned above, the components of the zircaloys have the following functions or effects. Sn is added for the purpose of improving the mechanical properties of the alloy and eliminating unfavorable effect on the corrosion resistance which may otherwise be caused by nitrogen contained in sponge zirconium used as a raw material for producing the zircalloys. Fe, Cr and Ni are added mainly for the purpose of improving the corrosion resistance. Discussion is made in the article as to the corrosion resistance in high temperature water of 315.degree. to 360.degree. C. and in steam of 400.degree. C. with respect to ternary alloys produced by adding a single element of Fe or Cr or Ni to each of Zr-2.5 wt % Sn alloy and Zr-1.8 wt % Sn alloy as well as binary alloys produced by adding a single element of Fe or Cr or Ni to Zr. The conclusion is that the optimum contents of Fe, Cr and Ni, when each of them is added as a single additive, are 0.22 wt %, 0.1 wt % and 0.22 wt %, respectively. Discussion is made also in regard to the effect of addition of Fe, Cr and Ni in combination. The article reports that the optimum total content of Fe, Cr and Ni is 0.35 wt % in a case of the steam of 400.degree. C. and is 0.3 wt % in another case of the water of 360.degree. C. The alloy compositions of the zircaloy-2 and zircaloy-4, which are presently used commonly, have been determined through the discussion explained above.
Thus, high levels of corrosion resistance of the zircaloy-2 and zircaloy-4 have been confirmed. However, ASTM, STP No. 633 (1977) pages 236-280 and pages 295-311 states that, when the zircaloy-2 and the zircaloy-4 with confirmed high corrosion resistance are used in a boiling water reactor, a papular local corrosion is observed to occur on the members made of these alloys. This local corrosion is generally known as nodular corrosion. As the high degree of burn-up of nuclear fuel is effected, areas suffering from the nodular corrosion are increased to connect one another and finally exfoliate from the material. Thus, the prevention of the nodular corrosion becomes essential to the operation of nuclear reactor with high degree of burn-up of the nuclear fuel.
ANS TRANSACTION Vol. 34 (June 1980) pages 237-238, J. Electrochem. Soc. Electrochemical Science and Technology, February 1975, pages 100-204, as well as Japanese Patent Laid-Open No. 95247/1983, state that the nodular corrosion which generally takes place in nuclear reactor can be well reproduced in an accelerated corrosion test conducted outside the reactor by using high temperature steam atmosphere of about 500.degree. C. or higher. In other words, it has been confirmed that the sensitivity of the zircaloy to the nodular corrosion cannot be evaluated through a test conducted in high temperature steam of 400.degree. C. or in high temperature water of 315.degree. to 360.degree. C. Corrosion test conducted under such an improved testing condition, i.e., within the atmosphere of high temperature steam of 500.degree. C. or higher, proved that even the zircaloys-2 and -4 are not sufficiently resistant to nodular corrosion. This in turn has given a rise to the demand for cladding tubes having higher resistance to nodular corrosion.
The specification of U.S. Pat. No. 2,772,964 discloses an alloy consisting of 0.1 to 2.5 wt % of Sn, not greater than 2 wt % of at least one of Fe, Cr and Ni, and the balance substantially Zr, but fails to disclose any alloy which is superior regarding both corrosion resistance and hydrogen absorption characteristics.
Japanese Unexamined Patent Publication Nos. 110411/1976, 110412/1976 and 22364/1983 disclose a heat-treating method known as .beta. quench for improving corrosion resistance of zircaloy, and also a process which comprises the .beta. quench step. Briefly, the .beta. quench method is a heat-treating method in which a zircaloy is quenched from a temperature range of .alpha.+.beta. phases or .beta.-phase alone. This treatment causes refining or partial solid-solution of intermetallic compound phases such as (Zr(Cr, Fe).sub.2, Zr.sub.2 (Ni, Fe), etc.) which are precipitated in the alloy. It is true that the .beta.-quenched zircaloy exhibits improved corrosion resistance, but the zircaloy of as .beta.-quenched state exhibits a low ductility due to the fact that it contains martensitic structure (acicular structure) which has super-saturated solid solution of Fe, Cr and Ni.
In order to improve the ductility of the zircaloy, therefore, it has been proposed to subject the zircaloy to a process in which a cold working and annealing are repeated alternatingly after the .beta. quenching, so as to obtain a recrystallized structure.
For instance, in the case of production of a nuclear fuel cladding tube, an ingot formed from a molten material is formed into a cylindrical billet through hot forging conducted at about 1000.degree. C., a solid-solution treatment conducted at about 1000.degree. C., hot forging conducted at about 700.degree. C. and hot extrusion. The billet is then subjected to .beta. quench followed by three repetitions of the alternating steps of Pilger mill cold rolling and annealing. If the steps of intensive working and annealing are repeated a plurality of times after the .beta. quenching, a coarse intermetallic compound phase will be caused in a zircaloy alloy having been improved to have high corrosion resistance by the .beta.-quenching, so that the corrosion resistance thereof becomes degraded.
Thus, it is desired that a zirconium based alloy used as a fuel cladding tube has a high corrosion resistance which does not vary when it is subjected to working and heat treatment.
The conventional methods described hereinabove for improving the corrosion resistance of zircaloy rely upon heat treatments, and no consideration has been made for the purpose of prevention of nodular corrosion through reconsideration of alloy composition. The conventional methods, therefore, could not completely prevent the nodular corrosion from occurring in a cladding tube used in the actual nuclear reactor. In addition, these known methods could not sufficiently reduce hydrogen absorption rate by the zircaloy.